1. Field of the Invention
This invention is concerned with a model steam generator having a system for facilitating the inspection of sample heat exchange tubes disposed within the generator.
2. Description of the Prior Art
Model steam generators for monitoring the amount of corrosion degradation occurring within the heat exchange tubes of a nuclear steam generator are known in the prior art. Generally speaking, such model generators operate by subjecting an array of sample heat exchange tubes in the same heat, pressure and chemical conditions which surround the heat exchange tubes in nuclear steam generators. If these conditions are accurately simulated, the amount of corrosion which occurs in the sample tubes of the model steam generator will provide an accurate indication of the tube corrosion present in the nuclear steam generator being monitored. Such model steam generators are a particularly useful form of corrosion monitor, because they obviate the need for shutting down the nuclear plant and sending technicians into the radioactive interiors of the generators. However, such model steam generators are useful only insofar as they are capable of accurately simulating the heat, pressure and chemical conditions which exist inside the nuclear plant. Any material departures from these conditions will adversely affect the accuracy of the model steam generator.
In order to understand the difficulties in building a practical model steam generator which provides an accurate monitor for heat exchange tube corrosion, one must first understand how nuclear steam generators are generally constructed, and what chemical and hydraulic conditions are responsible for tube corrosion.
Nuclear steam generators are comprised of three principal parts, including a secondary side and a tubesheet, as well as a primary side which circulates water heated from a nuclear reactor. The secondary side of the generator includes a plurality of U-shaped tubes, as well as an inlet for admitting a flow of feedwater. The inlet and outlet ends of the U-shaped tubes within the secondary side are mounted in the tubesheet which hydraulically separates the primary side of the generator from the secondary side. The primary side in turn includes a divider sheet which hydraulically isolates the inlet ends of the U-shaped tubes from the outlet ends. Hot water flowing from the nuclear reactor is admitted into the section of the primary side containing all of the inlet ends of the U-shaped tubes. This hot water flows through these inlets, up through the tubesheet, and circulates around the U-shaped tubes which extend within the secondary side of the steam generator. The heated water transfers its heat through the walls of the U-shaped tubes to the feedwater flowing through the secondary side of the generator, thereby converting the feedwater to steam. After the nuclear-heated water circulates through the U-shaped tubes, it flows back through the tubesheet, through the outlets of the U-shaped tubes, and into the outlet section of the primary side, where it is recirculated back to the nuclear reactor. The inlet ends of the U-shaped tubes are known as the "hot legs", and the outlet ends of these tubes are known as the "cold legs".
The heat exchange tubes of such nuclear steam generators can suffer a number of different types of corrosion degradation, including denting, stress corrosion cracking, intragranullar attack, and pitting. In situ examination of the tubes within these generators has revealed that most of this corrosion degradation occurs in what are known as the crevice regions of the generator. Such crevice regions include the annular space between the heat exchange tubes and the tubesheet, as well as the annular clearance between these tubes and the various support plates in the secondary side which are used to uniformly space and align these tubes. Corrosive sludge tends to collect within these crevices from the effects of gravity. Moreover, the relatively poor hydraulic circulation of the water in these regions tends to maintain the sludge in these crevices, and to create localized "hot spots" in the tubes adjacent the sludge. The heat radiating from these "hot spots" acts as a powerful catalyst in causing the exterior surface of the heat exchange tubes to chemically combine with the corrosive chemicals in the sludge. While most nuclear steam generators include blow-down systems for periodically sweeping the sludge out of the generator vessel, the sludges in the crevice regions are not easily swept away by the hydraulic currents induced by such systems. Despite the fact that the heat exchange tubes of such nuclear generators are typically formed from corrosion-resistant Inconel stainless steel, the combination of the localized regions of heat and corrosive sludges can ultimately cause the heat exchange tubes to crack, and leak radioactive water from the primary side into the secondary side of the generator. However, this need not occur if the heat exchange tubes are provided with internally reinforcing sleeves before the corrosion causes cracks in the tube walls.
Model steam generators were developed in order to accurately monitor the amount of corrosion degradation occurring in the heat exchange tubes of a particular nuclear steam generator, in order that these tubes might be sleeved before any of the tube walls crack. Such model steam generators have been found to be a particularly accurate way of ascertaining the amount of corrosion degradation occurring in the heat exchange tubes of a nuclear steam generator, because the particular amount of corrosion which the feedwater chemistry and thermohydraulics of the particular generator will induce in a particular set of tubes is virtually impossible to predict by purely theoretical models.
However, such prior art model steam generators are not without significant shortcomings. For example, in many of these model steam generators, the sample heat exchange tubes contained within the secondary side are not readily accessible. In model steam generators utilizing full-length U-shaped heat exchange tubes, the tubes and the tubesheet may be accessed only by disconnecting the large and heavy body of the secondary side from the tubesheet. The disconnection of the secondary side from the tubesheet is a lengthy and arduous process, requiring the detachment of a large number of nuts from their respective mounting studs. After all of these nuts have been removed, the large and heavy cylindrical body of the secondary side must be lifted off the tubesheet by a crane or a winch-type mechanism. If it is desired to completely remove the body of the secondary side from the tubesheet, over a hundred feet of vertical space is required. Additionally, if one wishes to make a close examination of the tubesheet itself, the complete disconnection of the tubesheet from both the primary and secondary sides of such a model steam generator is an even more arduous, difficult and time-consuming procedure than merely disconnecting and lifting the secondary side of the generator.
While there are prior art model steam generators which use sample tubes which are considerably shorter than the full-length U-shaped tubes used in the nuclear steam generators, the disconnection of the secondary sides of these generators from their respective tubesheets still requires removal of a great number of nuts engaged to a plurality of studs which uniformly circumscribe the joint between the tubesheet and the secondary side. The reassembly of these components requires the remounting of each of these units in a highly specific torquing sequence; thus both the disassembly and reassembly of the generator remains a difficult and arduous process. While the amount of vertical space required for disassembly is lessened by the use of shorter tubes, such model steam generators still require the use of some sort of crane or winching assembly in order to pull the body of the secondary side away from the tubesheet. In both the long and short tube cases, the use of such cranes or winches to reassemble the secondary side upon the tubesheet is a delicate and awkward process, as these two components must be positioned into precise alignment before they may be reconnected.
Clearly, a need exists for a system for conveniently disassembling a model steam generator in order that the sample tubes contained therein may be easily inspected and replaced. Ideally, such a system should provide a means for conveniently disconnecting the entire tubesheet of the model steam generator in order that the tubesheets of varying thicknesses may be conveniently interchanged within the model steam generator. Finally, it would be desirable if such a system included a means for easily and conveniently positioning and aligning the tubesheet and primary side of the model steam generator relative to the secondary side without the need for cranes or winches.